Preparation of 99Mo/99mTc generator based on alumina 99Mo-molybdate (VI) gel

In this work alumina ⁹⁹Mo-molybdate (VI) gel is evaluated as a column matrix for use in the preparation of small chromatographic column type ^99mTc generator. Alumina molybdate (VI) gel is prepared by dissolving inactive MoO₃ with aluminum foil in 5 M NaOH solution containing ⁹⁹Mo radiotracer. After complete dissolution, 0.5 H₂O₂ was added to the reaction mixture solution and acidified to pH 5.5 with concentrated HNO₃. The formed AlMo precipitate was washed with NaNO₃ solution, dried at 50 °C for 24 h and then packed in the form of a chromatographic column for elution of the generated ^99mTc radionuclide with physiological saline solution (0.9 % NaCl). Greater than 86 % of the generated ^99mTc activity is immediately and reproducibly eluted with passing 10 mL of the saline solution through 2.0 g of alumina ⁹⁹Mo-molybdate column bed at a flow rate of about 1.0 mL/min. The high radiochemical ≥98.6 % TcO₄ ^?, radionuclidic ≥99.90 % ^99mTc and chemical purities of the eluates satisfy the specifications for use in nuclear medicine.     

Desorption of99Mo from spent99Mo/99mTc generator

A simple method for desorption and purification of⁹⁹Mo from spent⁹⁹Mo/^99mTc generators is described. The alumina column was washed successively with 0.9% saline water, 35% H₂O₂, and then the⁹⁹Mo was eluted with 2M NH₄OH. Ammonia and residual H₂O₂ were removed by heating the eluate. Finally,⁹⁹Mo solution was passed through a 0.2 m membrane filter to remove precipitated aluminium hydroxide.     

Technetium-99m generator based on 12-molybdocerate —99Mo precipitate as column matrix

The preparation of insoluble 12-molybdocerate(IV) from⁹⁹Mo of low specific activity, produced by thermal neutron irradiation of MoO₃, is described. Samples of the material are dried at 50, 100 and 200°C and used as column matrices from which the generated^99mTc activity is periodically eluted with saline solution or saline solution containing 5·10^–5M K₂CrO₄ as an oxidant. The elution yields of^99mTc are high and reproducible (95–81%) with radionuclidic purity 99.98%. Both chemical and radiochemical purity (as TcO ₄ ^– ) of the eluates decrease with increasing drying temperature of the column matrix. Using chromated saline solution as eluent improves the radiochemical purity of the^99mTc eluate.     

Separation of99mTc from parent99Mo by solid-phase column extraction as a simple option for a new99mTc generator concept

Evidence is obtained to show that the liquidliquid extraction separation of^99mTc from⁹⁹Mo with methyl ethyl ketone, methyl propyl ketone and methyl isobutyl ketone can be transformed into a solid-phase column extraction procedure. The aqueous alkaline⁹⁹molybdate solution is immobilized on a column of a granular large-pore diatomaceous earch support, which is the neluted with the abovementioned extractants. Rapid and clean separation of^99mTc can be with all three solvents. The^99mTc can be back-extracted from the organic phase on a column filled with distilled water /or saline/ loaded granular diatomaceous earth /Extrelut^®/. The possibility of using the abovementioned procedure as a basis for a new^99mTc/⁹⁹Mo generator concept is envisaged.     

A novel technique for the separation of99mTc from99Mo

A new technique for the separation of^99mTc from low specific activity⁹⁹Mo is reported. A separation based on the principle of precipitation of⁹⁹Mo as calcium molybdate has been investigated. On precipitating⁹⁹MoO ₄ ^2– from alkaline solution as calcium molybdate under controlled conditions, the^99mTcO ₄ ^– is found to remain quantitatively in the supernatant solution with little carry-over of⁹⁹Mo. This calcium molybdate (⁹⁹Mo) could be redissolved and reprecipitated at regular intervals, yielding^99mTc quantitatively in aqueous neutral solutions. Calcium molybdate precipitates containing up to 1.5 GBq of⁹⁹Mo and 130–180 mg of molybdenum were prepared and evaluated. The performance in terms of repeated^99mTc separation gave yields of 75–93% with acceptable readionuclidic and radiochemical purity.     

The study of conditions for the preparation and application of99Mo−99mTc generators starting from irradiated molybdenum metal

⁹⁹Mo−^99mTc generators were prepared starting from irradiated molybdenum metal instead of MoO₃ in order to use reactor irradiation space more economically. The adsorption of molybdenum as sodium molybdate on different kinds of alumina was investigated. The effect of the pH of the column and the aqueous phase concentration of molybdenum were studied and related to the elution yield of^99mTc. A study of the radiation damage effect indicated that generators having a high elution efficiency of 90% could be prepared in the 100–600 mCi range. The losses of⁹⁹Mo were minimized to 10⁻⁵-10⁻⁴% and those of alumina to 2–5 μg/ml eluate.     

Pharmaceutical grade sodium [99mTc] pertechnetate from low specific activity 99Mo using an automated 99Mo/99mTc-TCM-autosolex generator

Performance study of a computer controlled automated closed cyclic module for the separation and recovery of ^99mTc from low specific activity (n, γ) ⁹⁹Mo using methyl ethyl ketone (MEK) solvent extraction technique named ⁹⁹Mo/^99mTc-TCM-AUTOSOLEX (Technetium automated solvent extraction) Generator is described. The entire system is automated and controlled by a user-friendly PC based graphical user interface that actually supervises process via an embedded system based electronic controller. The average yield of separation of ^99mTc was above 85 % and ⁹⁹Mo breakthrough in ^99mTc pertechnetate was <0.002 %. The sodium pertechnetate obtained was a clear solution having pH 6–7, Radiochemical (RC). Purity >99 %, MEK content <0.1 % (v/v), Al and Mo content <10 µg/ml. R. C. Purity of ^99mTc-radiopharmaceuticals studied was not less than 96 %. Bio-Quality control studies confirm that sodium pertechnetate obtained was sterile and pyrogen free. Imaging studies in animals and humans with limited radiopharmaceuticals show that the quality of ^99mTc-pertechenate obtained in the present module was good enough to do clinical study.     

Practicality of Tetragonal Nano-Zirconia as a Prospective Sorbent in the Preparation of 99Mo/99mTc Generator for Biomedical Applications

The feasibility of using tetragonal nano-zirconia (t-ZrO₂) as an effective sorbent for developing a ⁹⁹Mo/^99mTc chromatographic generator was demonstrated. The structural characteristics of the sorbent matrix were investigated by different analytical techniques such as XRD, BET surface area analysis, FT-IR, TEM etc. The material synthesized was nanocrystalline, in tetragonal phase with an average particle size of ~7 nm and a large surface area of 340 m² g^?1. The equilibrium sorption capacity of t-ZrO₂ is >250 mg Mo g^?1. The present study indicates that ⁹⁹Mo is both strongly and selectively retained by t-ZrO₂ at acidic pH and ^99mTc could be readily eluted from it, using 0.9% NaCl solution. A 9.25 GBq (250 mCi) t-ZrO₂ based chromatographic ⁹⁹Mo/^99mTc generator was developed and its performance was repeatedly evaluated for 10 days. ^99mTc could be eluted with >85% yield having acceptable radionuclidic, radiochemical and chemical purity for clinical applications. The compatibility of the product in the preparation of ^99mTc labeled formulations such as ^99mTc-EC and ^99mTc-DMSA was evaluated and found to be satisfactory.     

Rapid and sensitive determination of pertechnetate in99Mo/99mTc generator eluates by reversed-phase high-performance liquid chromatography

Reversed-phase high-performance liquid chromatography with u.v. detection was applied for rapid and sensitive determination of pertechnetate in⁹⁹Mo/^99mTc generator eluates, using a mixture solvent of acetonitrile and 0.04M aqueous acetate buffer (1/1) containing a few volume percentage of 0.5 M tetra-n-butylammonium hydroxide as the mobile phase. Employing a -bondapak C₁₃ column, the TcO ₄ ^– species was separated, monitored with absorbance at 254 nm, and observed at the retention time of 3.5 min. The detection limit was found to be 5.2·10^–10 g of Tc for each injection. Total Tc contents in the^99mTc eluates from clinically-used⁹⁹Mo/^99mTc generator were analyzed by this technique. The^99mTc (⁹⁹Tc) species was separated from the contaminant⁹⁹Mo. This method was found to be useful for the purification of^99mTc (⁹⁹Tc) as well as the determination of total Tc content.     

A99Mo-99mTc generator based on the use of zirconium molybdophosphate-99Mo gel

A modified⁹⁹Mo–^99mTc gel generator is described. The present generator uses an insoluble zirconium molybdophosphate (ZrMP) gel tagged with⁹⁹Mo. Molybdenum-99 is chemically combined in the gel structure and cannot be eluted from the matrix. The presence of phosphate increases the chemical stability of the gel and decreases the molybdenum breakthrough. The prepared gel is sufficiently porous to permit ready diffusion of^99mTcO ₄ ^– which can be cluted with saline in yields of up to 90%. The gel was found to contain 25.1% Mo, 21.9% Zr, and 0.7% P in a molar ratio of 1.09:1.0:0.09, respectively. The high molybdenum content of the gel allows the use of cheap, non-polluting (n, )⁹⁹Mo. The eluted^99mTc was of high purity and can be used for medical and pharmaceutical applications.     

Preparation of99Mo,132Te isotopes and99mTc,132I generators

Adsorption and desorption of⁹⁵Zr−⁹⁵Nb,⁹⁹Mo,¹⁰³Ru,¹³²Te and²³⁹Np in a HCl-alumina system were studied in order to purify⁹⁹Mo and¹³²Te obtained by the cation-exchange separation of fission products and to prepare highly pure^99mTc and¹³²I generators.⁹⁹Mo and¹³²Te, of which radionuclidic purity was over 99.99% and 99.999%, respectively, could be obtained by passing the cation-exchange separated Mo and Te fractions through alumina columns, by washing with HCl and finally by eluting⁹⁹Mo with 1M NH₄OH and¹³²Te with 3M NaOH. In order to raise the recovery of⁹⁹Mo and¹³²Te from the alumina columns, they should be eluted as quickly as possible after the adsorption. The direct use of the alumina column containing⁹⁹Mo or¹³²Te as the generator allowed milking of^99mTc or¹³²I, of which radionuclidic purity was over 99.999%. Milking yields of^99mTc with 0.1M HCl and¹³²I with 0.01M NH₄OH were 77% and 90%, respectively. The latter value was much higher than that in usual performance of the generator.     

<Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generators based on aluminum molybdate gel matrix prepared by nano method

A procedure for preparation of ⁹⁹Mo/^99mTc radioisotope generator based on low specific activity neutron activated ⁹⁹Mo was developed. Aluminum molybdate(VI)-⁹⁹Mo of high Mo(VI) content (~?364 mg/g Al⁹⁹Mo) was prepared by mixing low specific activity molybdate(VI)-⁹⁹Mo and aluminum mixture solution with isoamyl alcohol. Al⁹⁹Mo gel matrix was precipitated when the pH of the mixture solution was raised to ~?5 by addition of NaOH to the mixture. Radiometric measurements indicate the strong fixation of Molybdate(VI)-⁹⁹Mo species in the form of the sparingly insoluble Al⁹⁹Mo gel matrix. The prepared AlMo gel matrix was physiochemically characterized. Al⁹⁹Mo gel matrix was used as a base material for preparation of ⁹⁹Mo/^99mTc generator. The ^99mTc eluted from ⁹⁹Mo/^99mTc radioisotope generator was found to have relatively high elution yield (84?±?2.3%), radionuclidic (≥?99.99%), radiochemical (98.1?±?0.9%) and chemical purity.     

Radiochemical separation of99Mo from natural uranium irradiated with thermal neutrons

⁹⁹Mo was separated from uranium and insoluble fission product hydroxides. More than 98% of⁹⁹Mo radioactivity was extracted with bis (2-ethylhexyl)phosphoric acid. The organic phase was washed and⁹⁹Mo was back-extracted from the organic phase with NH₄OH solution. The percent recovery from the organic phase was 91% and the purity of⁹⁹Mo was more than 99%. Pure^99mTc was also extracted from the organic phase with a saline solution. Reversed-phase partition chromatography was used for the purification of⁹⁹Mo from¹³¹I and other fission products (10% HDEHP on kieselguhr bed).¹³¹I and other isotopes were quantitatively eluted with 0.1M H₂SO₄,⁹⁹Mo was eluted using a mixture of 0.5 M HCl and 30% H₂O₂.     

A novel electrochemical <Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generator

A novel electrochemical process to avail clinical grade ^99mTc from (n,γ)⁹⁹Mo has been demonstrated. The electrochemical parameters were optimized to maximize the ^99mTc yield with minimal ⁹⁹Mo contamination. ⁹⁹Mo/^99mTc generators containing up to 29.6 GBq (800 mCi) ⁹⁹Mo were developed and their performance were extensively evaluated for 10 days without changing the operating conditions. Very high radioactive concentration of ^99mTcO₄ ⁻ of acceptable quality, commensurate with hospital radiopharmacy requirements could be availed from the system with >90% yield. The compatibility of the product for the formulation of ^99mTc labeled radiopharmaceuticals such as ^99mTc-DMSA and ^99mTc-EC was found to be satisfactory in terms of high labeling yields. The proposed route represents an important step for enhancing the scope of accessing clinical grade ^99mTc from low specific activity (n, γ)⁹⁹Mo.     

Technetium-99m generators based on neutron irradiated 12-molybdocerate as column matrix

Neutron irradiated 12-molybdocerate(IV) is evaluated as a column matrix for use in preparation of small chromatographic column type^99mTc generators. Greater than 87% of the generated^99mTc activity is immediately and reproducibly eluted with passing 10 cm³ of saline or chromated saline solutions through 3.0 g of molybdocerate column bed at flow rates of about 1.0 cm³/min. The elution performance of^99mTc does not drastically change with increasing the drying temperature and/or the particles sixe of the column matrix in the range investigated. Saline eluent containing 5·10^–5M CrO ₄ ^2– increases the radiochemical purity of the eluate to >98% TcO ₄ ^– . Radionuclidic and chemical purities of the eluates satisfy the specifications for use in nuclear medicine.     

Separation of99mTc from neutron-irradiated MoO3 by precipitation as CaMoO4

A method is discussed for separating^99mTc from⁹⁹Mo by dissolution and reprecipitation of CaMoO₄ containing⁹⁹Mo-^99mTc.^99mTc can be obtained successfully with a yield of about 80% and with a radionuclidic purity as high as over 99.7%. On the other hand, by agitating Ca⁹⁹ MoO₄ solid with 0.9% NaCl aqueous solution,^99mTc can directly be obtained with a yield of about 30% without any impurities.     

Evaluation of proton induced reactions on 100Mo: New cross sections for production of 99 mTc and 99Mo

The use of the ⁹⁹Mo^99mTc generator in nuclear medicine is well established world wide. The production of the ⁹⁹Mo (T_1/2 = 66 h) parent as a fission product of ²³⁵U is largely based on the use of reactor technology. From the early 1990's accelerator based production methods to provide either direct produced ^99mTc or the parent ⁹⁹Mo, were studied and suggested as potential alternatives to the reactor based production of ⁹⁹Mo. A possible pathway for the charged particle production of ^99mTc and ⁹⁹Mo is irradiation of molybdenum metal with protons via the reaction ¹⁰⁰Mo(p,2n)^99mTc and ¹⁰⁰Mo(p,pn)⁹⁹Mo, respectively. The earlier published excitation functions show large differences in their maximum that result in large differences in the calculated yields. We therefore decided to study the excitation function for these proton-induced reactions. In this work the newly measured excitation functions as well as an evaluation of earlier measured data and a discussion of the observed disagreements are presented.     

Chemical and radiochemical evaluation of the purity of99mTc extracted by MEK

Solvent extraction separation of^99mTc from⁹⁹Mo using methyl ethyl ketone(MEK) has been found to be an effective method of obtaining^99mTc of medicinal purity from low specific activity⁹⁹Mo. The authors have investigated the effect of alkali and molybdenum concentration on the extraction of⁹⁹Mo and^99mTc into methyl ethyl ketone. The possibility of methyl ethyl ketone forming enol and condensation products and its effect on the final extraction efficiency and purity of^99mTc has been studied. Sodium molybdate has been found to have a good salting out effect on^99mTc pertechnetate and hence^99mTc extraction can be better accomplished from low specific activity⁹⁹Mo solutions. The ketone seems to form traces of condensation products in the extraction procedure. These have been found to be coextracted with^99mTc into MEK but did not affect the extractability of^99mTc. It was observed that neutral alumina column removes these condensation products from MEK containing^99mTc. Alternately these could be filtered off by acidification of the final aqueous^99mTc solution. The studies indicate that under optimum experimental conditions methyl ethyl ketone separates^99mTc from⁹⁹Mo with high efficiency and yields^99mTc of high purity suitable for use in nuclear medicine in the form of various labelled compounds.     

Separation of99Mo from132Te using thiourea as complexing agent. Application to the separation of99Mo from fission products

A radiochemical method to isolate⁹⁹Mo from¹³²Te, both produced in the fission of²³⁵U, has been developed. The method is based on the formation of a cationic complex of tellurium with thiourea in acid medium which is retained (98.7±0.5)% on a cation exchange resin (Dowex 50W-X8, 100–200 mesh), while (99.8±0.05)%⁹⁹Mo passes through it, due to the non-formation of such complex in the same experimental conditions. The radionuclidic purity of⁹⁹Mo was found to be suitable for the preparation of⁹⁹Mo–^99mTc generators. The retention of⁹⁹Mo on an alumina column as a function of pH was investigated and the best pH range for this purpose was found to be 4.0–4.5.     

Solid targets for 99mTc production on medical cyclotrons

Recent disruptions in the molybdenum-technetium generator supply chain prompted a review of non-reactor based production methods for both ⁹⁹Mo and ^99mTc. Small medical cyclotrons (E _p ~ 16–24 MeV) are capable of producing Curie quantities of ^99mTc from isotopically enriched ¹⁰⁰Mo using the ¹⁰⁰Mo(p,2n)^99mTc reaction. Unlike most other metallic target materials for routine production of medical radioisotopes, molybdenum cannot be deposited by reductive electroplating from aqueous salt solutions. To overcome this issue, we developed a new process for solid molybdenum targets based on the electrophoretic deposition of fine ¹⁰⁰Mo powder onto a tantalum plate, followed by high temperature sintering. The targets obtained were mechanically robust and thermally stable when irradiated with protons at high power density.